Neutronic analysis of hydride fueled PWR cores

The neutronic properties of U-ZrH_1.6 fuelled PWR cores are investigated and compared against those of the currently used UO₂ fuelled cores. In the first part of this work a parametric study is performed to quantify the neutronically achievable burnup for both hydride and oxide fuels at a number of enrichment levels and for a large number of geometries covering a wide design space of fuel rod outer diameter, D, and lattice pitch, P. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are calculated for hydride fuel with 5% and 12.5% enriched uranium. For this purpose a simplified procedure was developed that can, using single unit cell or assembly calculations, (1) account for non-linear burnup dependent k_∞ and thus to adequately predict the discharge burnup; (2) estimate the burnup dependent soluble boron concentration and; (3) estimate the reactivity coefficients; all of the above for a multi-batch core. In the second part of this work a detailed neutronic analysis is carried out for the six most economical geometries of both oxide and hydride fuels, with the purpose of designing the U-ZrH_1.6 fueled PWR cores to have negative reactivity coefficients. The preferred design found is replacement of 25 ^v/_o of the ZrH_1.6 by thorium hydride, along with addition of some IFBA burnable poison. It is also found that the conversion from oxide to hydride fueled PWR cores could be done without modifications in the control system.     

Neutronic design of hydride fueled BWRs

This paper summarizes the neutronic part of a study of the feasibility of designing BWR cores to have enhanced power density and simplified fuel bundle by using hydride instead of oxide fuel. A 3D fuel bundle neutronic analysis is performed for a limited number of geometries to determine attainable discharge burnup, pin-by-pin power distribution, axial power distribution, reactivity coefficients, reactivity worth of control elements and burnable absorber effects. It is found that hydride fuel bundle design can be simplified by eliminating water rods and partial length fuel rods and by reducing the volume of water in-between the fuel bundles. Both an ideal and more practical bundle designs are examined. A companion study of the thermal-hydraulic and vibration characteristics of BWR cores predicts that the increase in the number of fuel rods per given core volume enables increasing the BWR power density by up to ∼30% relative to oxide fuelled core design. The net outcome is expected to be improved BWR economics even though hydride fuel requires higher uranium enrichment to compensate for its reduced uranium loading.     

Economic analysis of hydride fueled BWR

The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 × 10 hydride fuel bundles instead of 10 × 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the “per volume base” fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Thermal-hydraulic analysis for wire-wrapped PWR cores

This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH_1.6 (referred to as U-ZrH_1.6) or UO₂ fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA).The thermal-hydraulic performance of U-ZrH_1.6 and UO₂ were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.     

Distribution of Plutonium Isotopes in Cooling Water from a PWR

Abstract

An investigation of plutonium isotopes in the primary cooling system of the Ringhals unit 2 (PWR) during normal operation has shown average concentrations of 0.03 Bq·l ^?1 of ²²⁸Pu and 0.02 Bq·l ^?1of ^{239 240}Pu. A major fraction of plutonium is associated with particles in contrast to dissolved plutonium in ionic form. The observed concentrations of plutonium isotopes in cooling water are characterized by log-normal distributions. The overall efficiency of the ion- exchange cleaning system with respect to plutonium in the primary circuit has been estimated as approximately 60%. During normal operation the ion-exchange cleaning system annually collects about 3 MBq of ²³⁸Pu and ²³⁹MBq of ^{239 +240}Pu. The cleaning system efficiently removes plutonium particles with sizes in the range 0.05 to 0.15 um from the primary cooling water. Particles with sizes outside this range are removed less efficiently.     

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.     

Reactivity feedback calculation of a conceptual TRISO fueled compact PWR core

Reactivity feedback coefficients have been calculated for a compact sized PWR core that utilizes carbon coated micro fuel particles instead of standard cylindrical fuel pellets with an inventive composition. A small amount of Pu-240 with 5 w/o has also been added in tristructural-isotropic (TRISO) fuel in place of U-238 for the reduction of excess reactivity. The values of fuel, moderator and void reactivity coefficients have been calculated at the middle of fuel cycle. All the reactivity coefficients were found negative which meet the design safety criteria. It was also observed that all reactivity feedback coefficients are interlinked and their effects are pronounced when coupled together.     

Economic analysis of grid and wire wrap supported hydride and oxide fueled pressurized water reactors

An economic analysis is performed to calculate the levelized unit cost of electricity (COE) for a pressurized water reactor (PWR) retrofitted with a range of potential U (45 wt.%)-ZrH_1.6 hydride and UO₂ oxide fueled geometries (i.e., combinations of rod diameter and pitch) supported by traditional grid spacers (square array) and wire wrap spacers (hexagonal array). The time frame considered in computing the COE is the remaining plant life, beginning at the time of retrofit. The goals of the analysis are twofold: (1) comparing the economic performance of UO₂ and U-ZrH_1.6 fuels for a range of retrofitted geometries supported by grid and wire wrap spacers; and (2) investigating the potential economic benefits for nuclear utilities considering retrofitting new fuels and/or geometries into existing PWR pressure vessels. Fuel cycle, operations and maintenance (O & M), and capital costs are considered.The economic performance of U-ZrH_1.6 and UO₂ fuels is found to be similar, with UO₂ fueled designs providing a slight advantage when supported by grid spacers, and U-ZrH_1.6 providing a slight advantage when supported by wire wrap spacers. These small differences in cost, however, are within the bounds of uncertainty of this study and are not believed to provide a strong economic argument for the use of one fuel type over the other.To demonstrate the potential economic benefits of retrofitted designs to nuclear utilities, two different comparisons are made. The first compares the COE for retrofitted designs with the COE for a reference PWR, assumed to have operated long enough to recuperate its initial capital investment. The costs for this reference PWR reflect the “do-nothing” case for current plant owners whose primary expenditures are fuel cycle and O & M costs. The second comparison introduces a different reference PWR that includes the costs to operate an existing unit and the cost to purchase power from a newly constructed PWR, for comparison with retrofitted designs which offer increased power relative to existing commercial PWRs.For the first comparison, no grid supported designs and only one wire wrap supported design (i.e., U-ZrH_1.6 Stretch Case) provide a lower levelized unit cost of electricity than the reference “do-nothing” PWR. The primary cause of this conclusion is the capital costs incurred by retrofitted designs to change the core geometry and, for many designs, to upgrade primary and secondary loop components for operation at higher power than the reference PWR. The reference “do-nothing” PWR cost in this first comparison includes only operations and maintenance as well as fuel cycle costs but does not include a capital component. For the second comparison, significant cost savings are demonstrated for both grid (15-19% savings) and wire wrap (30-40% savings) supported designs using U-ZrH_1.6 and UO₂ fuels. These cost savings are enabled by enhancing the pumping capacity of the primary system and, for wire wrap supported designs, by taking advantage of enhanced critical heat flux performance. The optimal geometry for retrofitted UO₂ and U-ZrH_1.6 fueled PWR cores supported by grid spacers is D_rod = 6.5 mm and P/D = 1.39. The cost savings over the second case reference PWR are ∼19 and 15%, respectively. The cost savings for retrofitted PWRs that incorporate wire wrap spacing are even larger because of operation at even higher power. Cost savings over the reference PWR range between 30 and 40% for the U-ZrH_1.6 and UO₂ Achievable and Stretch Cases. The optimal geometries for the U-ZrH_1.6 Achievable and Stretch Cases are D_rod = 8.08 mm, P/D = 1.41 and D_rod = 8.71 mm, P/D = 1.39, respectively. The optimal geometries for the UO₂ Achievable and Stretch Cases are D_rod = 7.13 mm, P/D = 1.42 and D_rod = 9.34 mm, P/D = 1.27, respectively. Utilities seeking to meet rising demand by expanding capacity may therefore strongly benefit from retrofitting existing PWRs with either U-ZrH_1.6 or UO₂ fueled designs. These new designs have different geometries than are currently used by commercial plants. A conclusion on which fuel type to use, however, could not be reached in this analysis as both offer similar economic performance.     

Calculation of the ex-core neutron noise induced by individual fuel assembly vibrations in two PWR cores

Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor(PWR) cores has been conducted to investigate the effect of cycle burnup on the properties of the ex-core detector noise. An extension of the method and the computational models of a previous work have been applied to two different PWR cores to examine a hypothesis that fuel assembly vibrations cause the corresponding peak in the auto power spectral density(APSD) increase during the cycle. Stochastic vibrations along a random two-dimensional trajectory of individual fuel assemblies were assumed to occur at different locations in the cores. Two models regarding the displacement amplitude of the vibrating assembly have been considered to determine the noise source. Then, the APSD of the ex-core detector noise was evaluated at three burnup steps. The results show that there is no monotonic tendency of the change in the APSD of ex-core detector; however, the increase in APSD occurs predominantly for peripheral assemblies. When assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core, the effect of the peripheral assemblies dominates the ex-core neutron noise.This behaviour was found similar in both cores.     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nucleate Boiling Ratio (DNBR) which utlimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithmFirst, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for these channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum.Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties.Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code.     

Radiochemical Determination of Plutonium Isotopes and Americium-241 in a Plutonium Sample

Isotopic ratios of ²³⁸Pu, ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Am in a plutonium sample were approximately determined by means of combined α- and γ-ray spectrometry without chemical separation of americium from plutonium. The intensity of α-ray followed by internal conversion was determined by measuring the intensity of LX-rays. The ratio of ²⁴⁰Pu to ²³⁹Pu was obtained from the ratio between the α-ray intensity thus determined and the total α-ray intensity from the two nuclides.     

热表面电离质谱法测定钚氧化物中的钚同位素组成

采用热表面电离质谱法对钚氧化物中钚同位素丰度进行了测定。通过对钚氧化物样品预处理、离子源和分析器的真空控制、法拉第杯接收效率检测、测量过程中的信号强度大小控制、信号强度稳定性控制以及测量时间的控制等条件进行优化,确定了最佳预处理条件和测量条件,实现了钚氧化物中钚同位素组分的准确测定。在选定的条件下,测定了钚标准样品中的钚同位素丰度,主同位素239Pu和242Pu测量精密度(sr)均优于0.05%(n=6)。     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

Plutonium recycle in heavy water-power reactors

The performance of a heavy-water reactor using U²³⁸ with an equilibrium concentration of U²³⁹ Pu²⁴⁰ Pu²⁴¹ nuclei and small additions of U²³⁵ and with natural and depleted uranium make-up is discussed. A portion of the spent fuel discharged from the reactor is cleaned up of fission fragments and recycled to the reactor. Another (smaller) portion is withdrawn from the cycle after plutonium has been extracted and is replaced in the core by natural uranium.Translated from Atomnaya Énergiya, Vol. 20, No. 1, pp. 26–29, January, 1966     

钚在粘土中的迁移实验

为了获取²³⁹Pu在粘土中的迁移规律,采用静态法和动态柱法实验对其在粘土的迁移进行了研究。结果表明,粘土对²³⁹Pu的吸附能力较强。在实际水流速为0.61cm/d条件下,²³⁹Pu经过150 d迁移距离小于0.2cm。与动态实验相比,采用分配系数计算延迟系数偏小,动态实验结果更接近实际。     

Plutonium rock-like fuel burning in LWR

Plutonium rock-like oxide(ROX) fuel burning in LWR has been studied. To improve reactivity insertion accident(RIA) behavior of zirconia(ZrO₂) type ROX(Zr-ROX) fuel PWR, small negative Doppler reactivity coefficient of the fuel is increased with the additives such as 24mol% ThO₂ or 15mol% UO₂ in the fuel. There is also an approach of a heterogeneous core with 1/3 ROX and 2/3 UO₂ fuels. From the loss of coolant accident(LOCA) analysis of Zr-ROX fuel PWR, the importance to decrease the large power peaking is shown. The ThO₂ additive can make it easier to flatten the power distribution in the core, and improve not only the reactivity accident behavior but also the LOCA behavior. The power flattening can also be achieved by reducing the content of Gd₂O₃ mixed in ZrO₂ and adding Er₂O₃ in place.

In the case of weapons-grade plutonium burning, the plutonium transmutation rate in Zr-ROX fuel LWR is about 0.9tonne/GWe/300 days, and far larger than that of full MOX LWR. The additives of ThO₂ or UO₂ decrease the plutonium transmutation rate, yet it is still larger than that in full MOX LWR by more than 2 times. Even in 1/3 Zr-ROX fuel core, the transmutation rate is comparable with the full MOX case. Total amount of discharged plutonium becomes less than 1/4 to 1/6 in these cores.